.sup.99m Tc is a widely used radionuclide in radiopharmaceutical and nuclear medicine applications. The particular medical advantage of this radionuclide is its very short half-life of about 6 hours. However, the short half-life creates manufacture and delivery problems, because the radionuclide must be used very soon after it is produced. For this reason, .sup.99m Tc is preferably supplied to hospitals on demand by an on-site generator, through disintegration of isotopic molybdenum (.sup.99 Mo) and chemical separation of the .sup.99m Tc product. High purity and high activity are important, so that the .sup.99m Tc product may be used immediately as a pertechnate, in the preparation of radionuclide tracer compounds, etc.
Current medical technology requires the use of radionuclide generators which can supply radioactive levels of at least 4 GBq, most often 8-12 GBq, and in some cases as high as 40 GBq per generator. Of radionuclide the generators now in use, elution generators are the most advantageous because they provide for rapid, efficient, and simple production of the desired radionuclides. However, in practice, most elution generators rely on aluminum oxide as a sorption material, which has a sorption capacity of only several percent by weight of molybdenum. This limits the activity of the generator to only several hundred MBq when natural isotopes of molybdenum are irradiated by a medium neutron flux ranging approximately between 10.sup.17 and 10.sup.18 n/m.sup.2 s. This level of activity is insufficient for medical applications. Known elution generators must therefore rely on either (a) the irradiation of enriched .sup.98 Mo as a target material for a high-intensity neutron flux; or (b) a carrier-free .sup.99 Mo isotope obtained by fission of uranium. These known devices and processes require a large capital investment; high energy and labor costs; a complex series of processing and purification steps involving highly radioactive components; the separation of .sup.99 Mo from uranium fission products, which are about twenty times more active than the useable radionuclide.
Thus, for many applications the advantages of known elution generators have been outweighed by the practical and economic disadvantages, and other types of generators have been sought. Methods of .sup.99m Tc production allowing its separation from low specific activity parent .sup.99 Mo are to be used. For example, a sublimation of .sup.99m Tc may be obtained from a suitable .sup.99 Mo compound. Or, .sup.99m Tc products may be extracted from a strongly alkaline aqueous solution of a molybdate of .sup.99 Mo by methylethylketone. These processes permit the production of .sup.99m Tc products of medically sufficient activity from low to medium neutron flux irradiation within the range of about 5.times.10.sup.16 to 5.times.10.sup.17 n/m.sup.2 s. However, these chemical methods are substantially more complex, time-consuming, and labor intensive than a common elution generator. They may not be economically and conveniently practiced within a self-contained on-site apparatus. Instead, a centralized manufacturing and processing center will generally supply technetium 99m produced by sublimation or extraction to local hospitals and clinics. Although miniaturized extraction and sublimation generators are available for on-site hospital use, they remain complex and expensive.
The technology of known .sup.99m Tc generators is discussed in R. E. Boyd, Recent Developments in Generators of .sup.99m Tc Radiopharmaceutical and Labeled Compounds, IAEA (Vienna: 1973), p. 1-26; R. E. Boyd, Technicium 99m Generators--The Available Options, Int. J. Appl. Rad. & Isot. (New York: 1982), Vol. 33, p. 801-2; and V. J. Molinsku, A Review of .sup.99m Tc Generator Technology, Int. J. Appl. Rad. & Isot. (New York: 1982), Vol. 33, p. 811-19.
Other practitioners have sought to improve elution generators by replacing the aluminum oxide sorption material with a sorption matrix. The matrix is intended to recover greater amounts of molybdenum from which .sup.99m Tc can be eluated, thus improving efficiency and yield. J. V. Evans, P. W. Moore, M. S. Shying, & J. M. Soddeau, "A New Generator For .sup.99m Tc," Third World Congress on Nuclear Medicine and Biology, pp. 1592-5 (Paris: 1982). The Evans device uses a sorption matrix of zirconium molybdate obtained from irradiated molybdenum oxide that is dissolved in a lye solution, precipitated by zirconium nitrate, and dried at 105.degree. C. The approximate chemical composition of this sorption material is ZrO.sub.2.MoO.sub.3.xH.sub.2 O, having a molybdenum concentration of approximately 25% by weight. Through a hydration and shaping process, the material achieves elution of .sup.99m Tc from .sup.99 Mo with an efficiency of 70-90%.
In experiments performed by the inventors in addition to zirconium molybdate also with titanium molybdate and polymolybdates of both elements containing 10-40% by weight molybdenum (with preferred content 20-30%) elution efficiencies 40-80% have been achieved. In contrary to previous authors (Evans et al) the elution matric was not made from already previously irradiated radioactive material dried at 105.degree. followed by hydration but in our case the elution matrix has been made from completely inactive material and dried at lower temperatures, prior to its activation by neutrons in the reactor. The drying has been performed at 40.degree.-50.degree. C., lasting many hours, in some cases even drying at room temperature (approximately. 20.degree. C.), lasting many days, has been used. The grain size of the matrix material usually has been in the range 50-140 /um. The inactive matrix prepared in such a way has been directly used as target material for the exposure to neutron activation in the reactor prior to elution.